1. Field of the Invention
The present invention relates generally to nuclear reactor fuel assemblies and, more particularly, is concerned with a boiling water nuclear reactor (BWR) fuel assembly incorporating fuel rod mini-bundles having reduced diameter interior fuel rods for improved critical heat flux (CHF) performance.
2. Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in heat exchange relation to the fuel rods so that the heat can be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of a self-sustained, nuclear fission reaction. The core is submersed in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and as a neutron moderator. Specifically, in a BWR the fuel assemblies are typically grouped in clusters of four with one control rod associated with each four assemblies. The control rod is insertable within the fuel assemblies for controlling the reactivity of the core. Each such cluster of four fuel assemblies surrounding a control rod is commonly referred to as a fuel cell of the reactor core.
A typical BWR fuel assembly in the cluster is ordinarily formed by a N by N array of the elongated fuel rods. The bundle of fuel rods are supported in laterally spaced-apart relation and encircled by an outer tubular channel having a generally rectangular cross-section. The outer flow channel extends along substantially the entire length of the fuel assembly and interconnects a top nozzle with a bottom nozzle. The bottom nozzle fits into the reactor core support plate and serves as an inlet for coolant flow into the outer channel of the fuel assembly. Coolant enters through the bottom nozzle and thereafter flows along the fuel rods removing energy from their heated surfaces. Such BWR fuel assembly is illustrated and described in U.S. Pat. No. 4,560,532 to Barry et al.
In a fuel assembly of this type the fuel rods in the central region of the bundle thereof may be undermoderated and overenriched. In order to remedy this condition by increasing the flow of moderator water through this region of the assembly, an elongated centrally-disposed water cross is frequently used in the assembly. The central water cross has a plurality of four radial panels which together form a cruciform water flow channel which divides the fuel assembly into four, separate elongated compartments, with the bundle of fuel rods being divided into mini-bundles disposed in the respective compartments. The water cross thus provides a centrally-disposed cross-shaped path for the flow of subcooled neutron moderator water within the channel along the lengths of, but separated from, adjacent fuel rods in the mini-bundles thereof. The fuel rods of each mini-bundle extend in laterally spaced apart relationship between an upper tie plate and a lower tie plate and are connected together with the tie plate to comprise a separate fuel rod subassembly within each of the compartments of the channel. The water cross has approximately the same axial length as the fuel rod subassemblies, extending between the upper and lower tie plates thereof.
Unlike other open lattice BWR fuel assembly designs, such as illustrated and described in U.S. Pat. Nos. 3,689,358 to Smith et al and 3,802,423 to Fritz et al, the above-described BWR of the Barry et al patent incorporates twice the amount of "cold or unheated wall" surface due to the presence of the water cross. Due to this aspect more of the liquid coolant tends to accumulate or cling onto the unheated part of the fuel assembly. This results in a relative starvation of liquid coolant at the heated surfaces in the mini-bundle interior locations.
CHF problems occur when the liquid film cooling a heated surface dries up. Such a coolant flow distribution would naturally tend to degrade substantially the CHF characteristics of this BWR fuel assembly design in its interior fuel rod locations. This thermal-hydraulic effect is clearly depicted in FIG. 4.42 on page 125 of a 1977 ANS Monograph entitled "The Thermal Hydraulics of a Boiling Water Nuclear Reactor" by R. T. Lahey et al. The figure shows the relative liquid/vapor distribution within a rod bundle assembly, where flow quality contours are depicted. Note that the lower the quality, the higher the liquid content, and vice versa. The tendency of liquid to cling on the walls and vapor to accumulate in the interior is clearly seen. Such a situation would cause premature dryout (i.e., at lower bundle power levels).
In the case of this BWR fuel assembly design, this reduction amounts to about 15-20% difference in performance for the corner/side versus interior rod locations based on recently completed CHF tests. Such a drastic variation essentially reduces the bundle critical power performance by up to 20% in cases of interior peaking. Clearly, a means of minimizing or eliminating this performance degradation is necessary.
Consequently, the need exists for further improvement of the BWR fuel assembly so as to prolong its useful life by improving significantly its CHF characteristics and performance.